This invention relates to austenitic stainless steel for use as a material for constituting a reactor such as a light water reactor, a fast breeder reactor and a nuclear fusion reactor, and more particularly to austenitic stainless steel having a superior irradiation-induced segregation resistance best suited for preventing the concentration change of the constituents in the crystal grain boundary (that is, for preventing irradiation-induced segregation) when the steel is exposed to a neutron-irradiating environment.
Conventionally, austenitic stainless steel SUS304 or SUS316 has been used as equipment parts of a reactor core of a light water reactor. During the operation of the nuclear reactor, the equipment parts of the reactor core is subjected to neutron irradiation, and therefore it is desired to use a material of a superior irradiation resistance for the equipment parts. Particularly, the most important properties with respect to the irradiation resistance of the material are irradiation embrittlement and stress corrosion cracking. It has been pointed out, for example, in Japanese Patent Unexamined Publication No. 57-155352 that the irradiation embrittlement can be improved by adding a trace amount of an additional element to the material. It is described in Japanese Patent Unexamined Publication No. 62-93075 that the stress corrosion cracking under irradiation can also be improved by adding a trace amount of an element.
In the above prior art for the present invention, however, the resistance to the irradiation embrittlement of the austenitic stainless steel due to the concentration change of the constituents in the crystal grain boundary which is induced upon receipt of high dose of neutron irradiation in the nuclear reactor, as well as the deterioration of the stress corrosion cracking resistance under irradiation, has not been recognized. Therefore, proper amount of the additive elements have been not studied.